ML15238A361
| ML15238A361 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/22/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML15238A360 | List: |
| References | |
| NUDOCS 9701240233 | |
| Download: ML15238A361 (4) | |
Text
A UNITED STATES 0
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON REACTOR COOLANT PUMP A2 MAIN FLANGE REPAIR THIRD TEN-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF NO. 96-05 OCONEE NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-269
1.0 INTRODUCTION
The Technical Specifications for Oconee Unit 1, state that the inservice inspection and testing of the American Society of Mechanical Engineers (ASME)
Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2 and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for Oconee Unit 1 for the current inservice inspection (ISI) interval is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose 9701240233 970122 PDR ADOCK 05000269 P
-2 alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
By letter dated December 26, 1996, Duke Power Company (licensee) requested relief from performing the Code-required volumetric examination of the repair to the Oconee Unit 1 reactor coolant pump main flange and proposed to use alternative examination which it was asserted, will provide an acceptable level of quality and safety.
The staff has reviewed and evaluated the licensee's request and the supporting information to use the proposed alternative to the Code-required examination of the weld repair to the Oconee Unit 1 reactor coolant pump main flange.
2.0 DISCUSSION Component Identification 1A2 Reactor Coolant Pump (RCP) Main Flange ASTM A351 (Steel Casting, Austenitic, for High Temperature Service) Grade CF8 Heat# 26554-1 Serial# 335.
Construction and Repair Code The Reactor Coolant Pump 1A2 was purchased to the ASME Boiler & Pressure Vessel Code,Section III, 1965 Edition with Addenda through summer 1967. The base metal indications were repaired under the rules of ASME Boiler & Pressure Vessel Code, Sections III and XI, 1989 Edition with no Addenda as required by the applicable inservice inspection code.
Code Examination Requirement The ASME Boiler and Pressure Vessel Code Section III, Subsection NB-2570 provides the non-destructive examination requirements for the repair of statically and centrifugally cast components. Paragraph NB-2571 requires radiographic examination (RT) after repairs.
Licensee's Basis for Requesting Relief (as stated)
In accordance with 10 CFR 50.55a(a)(3)(i), Duke Power proposes to use alternative examinations which will provide an acceptable level of quality and safety in lieu of the required ASME Section III radiography (RT) examinations. Furthermore, in accordance with 10 CFR 50.55a(a)(3)(ii), Duke Power believes that pulling the pump to enable the performance of the Code required RT examination would result in an excessive burden without a compensating increase in the level of quality or safety.
The alternative examination, layered liquid penetrant testing, in addition to the metallurgical root cause analysis and stress analysis (Attachment B, "Evaluation of Indications"), will provide a level of quality and safety equivalent to that obtainable through RT for a component of this size and material with the type of flaws identified.
-3 The hardship incurred due to outage extension and personnel radiation exposure is not justified by the limited information obtained from an RT. To gain sufficient access to the RCP main flange to perform an RT, the pump flange would have to be removed. Removal of the main flange requires removal of the pump motor and other interference items (piping, structural steel, hangers, etc.) as well as pump internals. This scope of work constitutes an outage evolution of three weeks. In addition, this work would impact the ability to perform the Oconee Emergency Power Engineered Safeguards Functional Test. This impact could result in an outage delay of approximately seven days. Exposure estimates from similar maintenance evolutions indicate that the expected personnel dose would be in excess of 2 rem.
Proposed Alternative Examination (as stated)
In lieu of RT, Duke Power's alternate examinations will consist of intermediate liquid penetrant examinations (PT) on all repair welding.
The intermediate PTs will be performed on all root passes and after each 1/2-inch of weld metal deposited. Additionally, the final weld PT examinations will be extended to include 2 inches of the surrounding base metal.
Lastly, following machining, a comprehensive final PT will be performed on accessible surfaces from approximately 3 inches outboard of the lower seal housing bolt holes into the flange bore.
Even if the pump were to be disassembled for RT examination of the pump main flange repaired areas, the flange.would require a radiography shot through approximately 14 inches of material (Reference Attachment A). A thickness of this magnitude greatly reduces the exam sensitivity (2 percent of 2T where T is the material thickness). The resulting minimum flaw size which could be detected by an RT exam is greater than 1/2 inch. Additionally, the subject base metal repairs lie in an area outside the minimum required coverage for the original manufacturer's RT in accordance with NB-2575.2-1(e) and note (c).
Ultrasonic (UT) examination is also impractical as an alternative to the RT due to the large grain size in the austenitic stainless steel casting of this type.
The intermediate layered PT examinations will effectively provide a volumetric exam of the weld repaired areas and ensure that there are no significant flaws in the welds. PT examination also has a high sensitivity and can identify indications that are difficult to detect visually. The comprehensive final PT will assure that no cracking has occurred adjacent to the weld repaired areas or other areas potentially prone to cracking. Therefore, the PT exams will provide adequate assurance of component integrity following repair.
3.0 EVALUATION AND CONCLUSION The staff has reviewed and evaluated the licensee's request for relief from performing the Code-required radiographic examination of the weld repair of the Unit 1 reactor coolant pump (A2) main flange. The licensee has proposed to perform the following alternatives.
Liquid penetrant examinations of root pass and each layer of one-half inch of weld metal followed by examination of cover pass.
-4 Metallurgical investigation for root cause of cracking.
Stress analysis of the main flange to determine maximum value of stress for use in fracture mechanics evaluation.
Fracture evaluation to determine sensitivity of the main flange to the presence of flaws.
The staff believes that the Code-required radiographic examination of the repaired areas lacks sensitivity to detect flaws smaller than one-half of an inch due to the magnitude of thickness being radiographed. Also, the other volumetric examination such as the ultrasonic examination of the flange as an alternative to radiographic examination, is impractical due to large grain size of cast stainless steel material as that of the main flange. However, the proposed liquid penetrant examination is a surface examination but very sensitive to detection of fine cracks such as the shrinkage cracks and the intergranular stress-corrosion cracks in stainless steel material.
When this examination is performed on layers of deposited weld metal, it is believed to provide adequate assurance of a sound weld metal.
Furthermore, the licensee performed a fatigue crack growth analysis and fracture evaluation to determine the sensitivity of the pump main flange to the presence of flaws. A conservative analysis of crack growth as a function of number of startup shutdown cycles for postulated flaws with depths from 0.5 to 2.5 inches indicate that the fatigue crack growth rate is negligible for all but the largest flaw which will even withstand three times the design number of cycles. Since the flange material has good fracture toughness and the thickness in the bore region where cracks originally occurred is 30 inches, there is a large margin against any potential failure in this region.
Therefore, the licensee's proposed alternative would provide an acceptable level of quality and safety and is authorized pursuant to 10 CFR 50.55a(a)(3)(i).
Principal Contributor:
Prakash Patnaik Date:
January 22, 1997